1. Field of the Invention
The present invention relates to methods and apparatus for measuring wear of control rods in nuclear reactors. More particularly, the invention is concerned with an eddy current measuring device which measures volume and thickness variations in the rod at various axial locations. Information thus generated is processed to produce a profile of the rod wear in order to determine available rod useful life and to remove from service rods exceeding the useful life limit.
2. Description of the Prior Art
In a typical nuclear reactor, the reactor core is composed of a plurality of elongated fuel assemblies, each of which contains a plurality of elongated fuel elements or rods. Liquid coolant is pumped upwardly through the core in order to extract heat generated in the core for the production of useful work. The heat output of the core may be regulated by the movement of control rods each comprising an outer tubular sheath containing neutron-absorbing materials therein.
Control rods may wear and fail for a variety of reasons. For example, during power operation of the reactor, most of the regulating control rods are maintained substantially withdrawn from the reactor core, except for a lowermost end of which is located within the upper ends of the guide thimbles. Withdrawn control rods are maintained in spatial relationship with each other by superadjacent guide structures which are known. When in such withdrawn positions, the control rods may experience significant vibration induced by coolant water flow within the guide thimbles and the guide structures. The vibration results in lateral contact of the rod against the internal wall surfaces of the guide thimbles, and the upper guide structures, etc. Consequently, the rod surfaces wear. Also, reciprocal movement of the control rods into and out of the fuel bundle results in axial wear. It is further possible that the control rod cladding surface may crack from other unspecified causes. Continuous wear of the rod cladding can lead to perforation of the rod and exposure of the neutron absorbing contents to the coolant/moderator fluid, which is undesirable.
In the past, attempts to measure, study and analyze rod defects have been conducted for the most part visually in the course of periodic refueling operations. Such attempts have been subject to high uncertainty because visual methods produce qualitative and somewhat ambiguous results. Non-visual methods of detecting defects in pipes have been known for some time. Such methods utilize magnetic leakage and eddy current techniques. However, these methods only sense gross defects such as cracks and deformities, but do not accurately and unambiguously determine the extent or degree of wear in a manner which is convenient to implement. Thus, especially in the case of older reactors, there has arisen a need to more accurately and unambiguously measure control rod wear.